Zirconium alloys have been used as nuclear fuel cladding material in fuel assemblies for nuclear power reactors due to their relatively low neutron cross section and high corrosion resistance. Two zirconium alloy groups, including the traditional Zircaloy-2™ in boiling water reactors and Zircaloy-4™ in pressurized water reactors (PWR) have been used as cladding material. Newer materials, such as ZIRLO™ (Zr-1Nb-1Sn-0.1Fe in wt %) and M5® (Zr-1Nb-0.04Fe in wt %) have also been evaluated. During reactor operations, the cladding typically undergoes outer surface corrosion as high temperature water reacts with the cladding producing hydrogen. A fraction of this hydrogen is then absorbed by the cladding. The total hydrogen concentration generally depends on temperature, fuel burn-up, and material type. The hydrogen concentration could be up to 600 ppm. J. P. Mardon et al, Update on the Development of Advanced Zirconium Alloys for PWR Fuel Rod Claddings, Proceedings of the 1997 International Topical Meeting on LWR Fuel Performance, Portland Oreg., La Grange Park, Ill.: American Nuclear Society, pp. 405-412.
During extended dry storage, cladding plays an important role in safely handling, storing, and transferring spent nuclear fuel. As the cladding cools with time during extended storage, the hydrogen inside the cladding may precipitate as hydrides because the solubility of hydrogen in zirconium decreases with temperature. Furthermore, both existing and newly formed hydrides may reorient. Depending on size, distribution, and orientation, these hydrides may induce premature fracture as a result of hydride embrittlement or delayed hydride cracking. Hydride embrittlement and reorientation of spent nuclear fuel cladding is therefore a potentially significant operational and safety concern.
As the industry has considered extended dry storage as an alternative approach to manage the spent nuclear fuel and the amount of high burn-up fuel is increasing as a result of changes in plant operating conditions, there is a need to evaluate the effects of hydriding on metallic substrate materials. See, e.g., R. L. Sindelar et al, Materials Aging Issues and Aging Management for Extended Storage and Transportation of Spent Nuclear Fuels, NUREG/CR-7116; SRNL-STI-2011-00005, Washington D.C.
Traditional methods of hydriding using electrochemical charging followed by annealing and charging in hydrogen gas are relatively time-consuming involving multiple steps where several days are required and the hydrides must be processed at high temperature (e.g., 300° C.).
Accelerated hydriding methods that can be operated at relatively lower temperature would be very important for selection and evaluation of metallic materials for cladding applications. In addition, hydriding at relatively lower temperatures could be extended to other applications such as evaluating hydrogen embrittlement of oil and gas pipeline materials.